This invention relates generally to the field of nuclear reactors and particularly to a method for improving the performance of irradiated structural materials.
Since the first commercial nuclear reactor produced steam for electric power in 1958, most reactors used have been based upon nuclear fission. Principally, light water reactors predominate in the U.S., heavy water reactors predominate in Canada, and gas-cooled reactors are used in the United Kingdom. The fast breeder reactor is in an advanced state of development in a number of countries including France, the U.K., West Germany, and Japan. Fusion reactors are still in the fairly early stages of development.
Despite the growing number of nuclear reactors and improvements in both the design and materials used in their construction, a major problem is the replacement and disposal of radiation damaged structural components. One of the most difficult materials problems is associated with the structural materials of the first wall of fusion reactors.
Most of the present fusion reactor design studies are based on the D-T fusion reaction. This reaction has the lowest ignition temperature and the lowest confinement requirements of all potential reactions. However, most of the energy comes out in the form of 14 MeV neutrons which cause damage to and induce radioactivity in the reactor's first wall structures. To help achieve economical fusion power, increased first wall lifetimes of up to 20-40 years are very desirable. Irradiation damage for a projected 20 year lifetime of the first wall is approximately 400 dpa and accumulated helium is of the order of 6000 appm for a reactor with 2 MW/m.sup.2 (d, t) neutron wall loading. There are no conventional alloys available at the present time which are known to be able to withstand these damage levels.
As a result, development of first wall alloy materials has focused on increasing the service life in the fusion environment of known materials. Examples of candidate alloy systems include austenitic stainless steels, ferritic/martensitic steels, and reactive and refractory alloys. These are being systematically studied under simulated fusion reactor conditions and are being optimized through microstructural and compositional modifications. However, even after optimization, the fusion reactor environment produces irreversible structural damage in these alloy systems, and, at a certain damage level, the first all components have to be replaced. Currently available data for all alloys which have been irradiation tested to date provide little hope than an exposure of 400 or more dpa, at the desired operating temperatures, will be possible with acceptable residual material properties. For example, austenitic stainless steels show a steady-state swelling of approximately 1%/dpa. Ferritic alloys appear to have a longer incubation period followed by a minimum swelling rate of approximately 0.06%/dpa.
It is therefore an object of the present invention to provide a process for increasing the lifetimes of nuclear reactor components and thereby increasing the efficiency and decreasing the cost of operating and maintaining the nuclear reactors.
It is a further object of the invention to provide a process for extending the lifetimes of components of both nuclear fission reactors and nuclear fusion reactors.
It is another object of the invention to provide a class of structural materials which can be used to fabricate nuclear reactor components with extended lifetimes despite high levels of irradiation.
It is a still further object of the invention to provide a process for in situ regeneration of nuclear reactor components.